GENERATION IV NUCLEAR ENERGY SYSTEMS
Transcript of GENERATION IV NUCLEAR ENERGY SYSTEMS
Brookhaven Science AssociatesU.S. Department of Energy
GENERATION IV NUCLEAR ENERGY SYSTEMSHOW THEY GOT HERE AND WHERE THEY ARE GOING
GENERATION IV NUCLEAR ENERGY SYSTEMSGENERATION IV NUCLEAR ENERGY SYSTEMSHOW THEY GOT HERE AND WHERE THEY ARE GOINGHOW THEY GOT HERE AND WHERE THEY ARE GOING
David J. DiamondBrookhaven National Laboratory
Energy Sciences and Technology DepartmentNuclear Energy and Infrastructure Systems Division
Presented at the University of TennesseeApril 30, 2003
Slide 2
OUTLINE OF PRESENTATIONOUTLINE OF PRESENTATION
n Introduction to the Gen IV (long-term) Nuclear Energy Systems
n The Roadmap - how we got to the Gen IV concepts
n The Not-Gen IV Nuclear Energy Systemsaka the international near-term deployment concepts
n What do the Gen IV concepts look like; what are some of their R&D needs
Slide 3
GENERATION IV NUCLEAR ENERGY SYSTEMSGENERATION IV NUCLEAR ENERGY SYSTEMS
Sodium Fast Reactor (SFR)
Gas-Cooled FastReactor (GFR)
Molten Salt Reactor(MSR)
Supercritical Water Reactor(SCWR)
Very High Temp.Gas Reactor (VHTR)
Lead-alloy Fast Reactor (LFR)
R&DApplicationsSizeFuel
CycleNeutron
Spectrum
Slide 4
GENERATION IV NUCLEAR ENERGY SYSTEMSGENERATION IV NUCLEAR ENERGY SYSTEMS
ClosedFastSodium Fast Reactor (SFR)
ClosedFastGas-Cooled FastReactor (GFR)
ClosedThermalMolten Salt Reactor(MSR)
Open,Closed
Thermal,Fast
Supercritical Water Reactor(SCWR)
OpenThermalVery High Temp.Gas Reactor (VHTR)
ClosedFastLead-alloy Fast Reactor (LFR)
R&DApplicationsSizeFuel
CycleNeutron
Spectrum
Slide 5
GENERATION IV NUCLEAR ENERGY SYSTEMSGENERATION IV NUCLEAR ENERGY SYSTEMS
Advanced Recycle
Electricity,Actinide Mgmt.
Med toLarge
ClosedFastSodium Fast Reactor (SFR)
Fuels, Materials,Safety
Electricity, Actinide Mgmt., Hydrogen
MedClosedFastGas-Cooled FastReactor (GFR)
Fuel, Fuel treatment,Materials, Safety and Reliability
Electricity, Actinide Mgmt., Hydrogen
LargeClosedThermalMolten Salt Reactor(MSR)
Materials, SafetyElectricityLargeOpen,Closed
Thermal,Fast
Supercritical Water Reactor(SCWR)
Fuels, Materials,H2 production
Electricity, Hydrogen,Process Heat
MedOpenThermalVery High Temp.Gas Reactor (VHTR)
Fuels, Materials compatibility
Electricity, Actinide Mgmt., Hydrogen
Small toLarge
ClosedFastLead-alloy Fast Reactor (LFR)
R&DApplicationsSizeFuel
CycleNeutron
Spectrum
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THE TECHNICAL ROADMAPTHE TECHNICAL ROADMAP
n Discusses the benefits, goals and challenges, and the importance of the fuel cycle
n Describes evaluation and selection process
n Introduces the six Generation IV systems chosen by the Generation IV International Forum
n Surveys system-specific R&D needs for all six systems
n Collects crosscutting R&D needs
n GIF countries will choose the systems they will work on
n Programs and projects will be founded on the R&D surveyed in theroadmap
n Information available at gif.inel.gov/roadmap/
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TECHNOLOGY GOALSTECHNOLOGY GOALS
n Sustainability1. Provide sustainable energy generation that meets clean air
objectives and promotes long-term availability of systems and effective fuel utilization for worldwide energy production
2. Minimize and manage their nuclear waste and notably reduce the long term stewardship burden, thereby improving protection for the public health and the environment
n Safety and reliability3. Operations will excel in safety and reliability
4. Will have a very low likelihood and degree of reactor core damage
5. Will eliminate the need for offsite emergency response
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TECHNOLOGY GOALSTECHNOLOGY GOALS
n Economics6. Will have a clear life-cycle cost advantage over other energy
sources
7. Will have a level of financial risk comparable to other energy projects
n Proliferation resistance and physical protection8. Will increase the assurance that they are a very unattractive and
the least desirable route for diversion or theft of weapons-usable materials, and provide increased physical protection against acts of terrorism
Slide 9
INTERNATIONAL NEAR-TERM DEPLOYMENT (1/2)INTERNATIONAL NEAR-TERM DEPLOYMENT (1/2)n Deployment by 2015
n Industry involvement
n Improvement over current advanced LWR performance
n Advanced Boiling Water Reactors • ABWR-II• ESBWR• SWR-1000• HC-BWR
n Modular High-Temperature Gas-Cooled Reactors• GT-MHR• PBMR
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INTERNATIONAL NEAR-TERM DEPLOYMENT (2/2) INTERNATIONAL NEAR-TERM DEPLOYMENT (2/2)
n Advanced Pressure Tube Reactor• ACR-700
n Advanced Pressurized Water Reactors• AP-600• AP-1000• APR-1400• APWR+• EPR
n Integral Primary System Reactors• CAREM• IMR• IRIS• SMART
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GEN IV NUCLEAR ENERGY SYSTEMSGEN IV NUCLEAR ENERGY SYSTEMS
n Very High Temp. Gas Reactor (VHTR)
n Gas-Cooled Fast Reactor (GFR)
n Supercritical Water Reactor (SCWR)
n Sodium Fast Reactor (SFR)
n Lead-alloy Fast Reactor (LFR)
n Molten Salt Reactor (MSR)
Slide 12
SEQUENCED DEVELOPMENT OF HIGH TEMPERATUREGAS COOLED NUCLEAR ENERGY SYSTEMSSEQUENCED DEVELOPMENT OF HIGH TEMPERATUREGAS COOLED NUCLEAR ENERGY SYSTEMS
PMRGFR
> 950°C for VHT heat process
Fast neutrons & integral fuel cycle for high sustainability
VHTR
Slide 13
VHTR FOR HYDROGEN PRODUCTIONVHTR FOR HYDROGEN PRODUCTION
n Hydrogen demand is already large and growing rapidly• Heavy-oil refining consumes 5% of natural gas for hydrogen production
n Energy security and environmental quality motivate hydrogen as an alternative to oil as a transportation fuel
• Zero-emissions • Distributed energy opportunity
n Water is the preferred hydrogen “fuel”
• Electrolysis using off-peak power
• High-temperature electrolysis• High-temperature
thermochemical water splitting
Slide 14
VERY HIGH TEMPERATURE REACTOR (VHTR)VERY HIGH TEMPERATURE REACTOR (VHTR)
Characteristics
• He coolant• 1000°C outlet temperature• Reactor coupled to H2production facility
• 600 MWth, nominally based on MHTGR
• Coated particle fuel, graphite block (or pebble?) core
Slide 15
Reactor Cavity Cooling System
Reactor Pressure Vessel
Control Rod Drive Stand Pipes
Power Conversion System Vessel
Floors Typical
Generator
Refueling Floor
Shutdown Cooling System Piping
Cross Vessel (Contains Hot & Cold Duct)
35m(115ft)
32m(105ft)
46m(151ft)
GT-MHR REACTOR BUILDINGGT-MHR REACTOR BUILDING
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Control Rod Drive Assembly
Refueling Stand Pipe
Control Rod Guide tubes
Cold leg Core Coolant Upper Plenum
Central Reflector Graphite
Annular shaped Active Core
Outer Side Reflector Graphite
Core Exit Hot Gas Plenum
Graphite Core Support Columns
Reactor Vessel
Upper Plenum Shroud
Shutdown Cooling System Module Hot Duct
Insulation Module
Cross Vessel Nipple
Hot Duct Structural Element
Metallic Core Support Structure
Core Inlet Flow
Core Outlet Flow
Insulation Layer for Metallic Core Support Plate
Upper Core Restraint Structure
Control Rods
7m(23 ft)
23.7m(78ft)
2.2m(7ft)
8.2m(27ft) Dia Vessel Flange
Upper plenum -hot plume
mixing - “LOF”
Core -depressurized
cooldown
Flow between hotter/ cooler
channels - “LOF”
Lower plenum - hot
jet mixing
Natural convection and thermal radiation
Core flow -normal operation
GT-MHRGT-MHR
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CORE FLOW ISSUES DURING NORMAL OPERATIONCORE FLOW ISSUES DURING NORMAL OPERATION
n Calculation of the coolant channel temperatures during normal operation• Significant local variations in power occur across the core due the non-uniform
location of the reflectors, control rods, and burnable poison assemblies and due to the fuel loading
• Power variations are amplified in the hot channels due to the buoyancy resistance• Therefore the coolant temperatures can vary by more than + or - 200°C from the
average
n Calculation of the core lower plenum flow mixing and pressure drop• Hot jet mixing, complex 3-dimensional flow around the core supports, and the flow
acceleration near the hot duct need to be calculated
n Calculation of the hot duct coolant mixing and insulation effectiveness• Permeation of the hot gas into the insulation is a concern • The entrance conditions are somewhat uncertain, but the flow must be well mixed by
the time it reaches the turbine
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Thermal-hydraulic Issues
n Mixing of the gases during bypass events
n Flow distributions among the recuperators and recuperatorefficiency
n Hot streaks at the turbine inlet
Generator
Thrust Bearing
Turbine
High Pressure Compressor
Recouperator
Recouperator
Low Pressure Compressor
Precooler/ Intercooler
Cold Gas to Reactor
Hot Gas from Reactor
PCS Vessel
34m(112ft)
8.2m(27ft) Dia. Vessel Flange
POWER CONVERSION UNITPOWER CONVERSION UNIT
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THERMAL-HYDRAULIC ISSUES -ACCIDENT CONDITIONSTHERMAL-HYDRAULIC ISSUES -ACCIDENT CONDITIONS
n Rejection of the heat by natural convection and thermal radiation from the reactor pressure vessel outer wall to the passive cooling system
• Local effects around the hot duct need to be considered• Some separate effects proof testing may be needed
n Reliability, robustness, and effectiveness of the Reactor Cavity Cooling System
n Flow through the core during a loss of circulation accident• Up flow in the hot channels and down flow in the cool channels results in hot
plumes in the upper plenum• The hot and cold channel flow distribution and the upper plenum mixing are
uncertain• Low Reynolds number flow with turbulent, transitional, and/or laminar flow,
buoyancy effects, and gas property variations
n Core cool down during a LOCA
n Air or water ingress during a LOCA
Slide 20
GAS-COOLED FAST REACTOR (GFR)GAS-COOLED FAST REACTOR (GFR)
Characteristics• He (or SC CO2) coolant, direct
cycle energy conversion• 850°C outlet temperature• 600 MWth/288 MWe• U-TRU ceramic fuel in coated
particle, dispersion, or homogeneous form
• Block, pebble, plate or pin core geometry
• Combined use of passive and active safety systems
• Closed fuel cycle system with full TRU recycle
• Direct Brayton cycle energy conversion
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ADVANTAGES OF GFRADVANTAGES OF GFR
n GFRs share the sustainability attributes of fast reactors• Effective fissioning of Pu and minor actinides• Ability to operate on wide range of fuel compositions (“dirty fuel”)• Capacity for breeding excess fissile material
n Advantages offered by use of He coolant • Ease of in-service inspection• Chemical inertness• Very small coolant void reactivity (<ßeff)• Potential for very high temperature and direct cycle conversion
n High temperature potential opens possibilities for new applications, including hydrogen production
Slide 22
GFR R&D NEEDSGFR R&D NEEDSn Safety case difficult with low thermal inertia and poor heat transfer
properties of coolant
• Reliance on active and “semi-passive” systems for decay heat removal• Passive reactivity shutdown is also targeted
n High actinide-density fuels capable of withstanding high temperature and fast fluence
• Modified coated particle or dispersion type fuels, e.g.,– (U,TRU)C/SiC– (U,TRU)N/TiN
• Fuel pins with high-temperature cladding, e.g., infiltrated kernel particle
n Core structural materials for high temperature and fast-neutron fluenceconditions (ceramics, composites, refractory alloys)
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FUEL / CORE CONFIGURATIONSFUEL / CORE CONFIGURATIONS
n GFR• Metal or ceramic matrix (similar to prismatic)• Pin, plate types (ceramic, metallic)• Pebble/particle
C o m p o s i t e C e r a m i c sF u e l E l e m e n t C o r e L a y - o u t
C o r e V e s s e l
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SCWR: GENERAL CHARACTERISTICSSCWR: GENERAL CHARACTERISTICS
n LWR operating at higher pressure (>22.1 MPa) and temperature (280-550°C)• Operating conditions with fossil plant experience
n Higher thermal efficiency (44% vs. 33%)
n No change of phase• Larger enthalpy rise in core• Lower flow rate (~10% of BWR)• Lower pumping power (smaller pumps)
n Simplified direct cycle system• No recirculation• Smaller reactor pressure vessel/containment
n Thermal or fast spectrum possible• Fuel cycle flexibility
Improved Economics
Slide 25
SCWR: OPTIMIZATION OF LWR TECHNOLOGY SCWR: OPTIMIZATION OF LWR TECHNOLOGY
Reactor
Turbine/Generator
Reactor
Turbine/Generator
ReactorSteam-water separation system
Recirculation system
Turbine/Generator
ReactorSteam-water separation system
Recirculation system
Turbine/Generator
Reactor
Steam generatorPressurizer
Turbine/Generator
Reactor
Steam generatorPressurizer
Turbine/Generator
Reactor
Turbine/Generator
Reactor
Turbine/Generator
ReactorSteam-water separation system
Recirculation system
Turbine/Generator
ReactorSteam-water separation system
Recirculation system
Turbine/Generator
Reactor
Steam generatorPressurizer
Turbine/Generator
Reactor
Steam generatorPressurizer
Turbine/Generator
Reactor
Turbine/Generator
Reactor
Turbine/Generator
ReactorSteam-water separation system
Recirculation system
Turbine/Generator
ReactorSteam-water separation system
Recirculation system
Turbine/Generator
Reactor
Steam generatorPressurizer
Turbine/Generator
Reactor
Steam generatorPressurizer
Turbine/Generator
Slide 26
SCWR: EFFECT OF SIMPLIFICATIONSCWR: EFFECT OF SIMPLIFICATION
Slide 27
Fuel assembly
SCWR NEUTRONIC DESIGNSCWR NEUTRONIC DESIGN
n Considerations with core design• Large change in density axially• Average coolant density higher than
in BWR• Downward flow in water rods• Other moderators (BeO, ZrH2)• Square or hexagonal geometry
Water rod
nSafety consideration•Rod ejection accident•Negative moderator reactivity coefficient
Slide 28
BASIC DATA - HEAT TRANSFERBASIC DATA - HEAT TRANSFER
Heat-transfer data at prototypical SCWR conditions (i.e., supercritical water, complex bundle geometry, high heat flux) are needed.
Single-phase heat transfer: Data exist for either simple round tubes and/or surrogate fluids. Existing SCW heat-transfer database and correlations are inconsistent.
Existing correlations and models for SCW heat transfer exhibit large discrepancies and diverging trends
Slide 29
CODES - NUMERICAL INSTABILITIESCODES - NUMERICAL INSTABILITIES
Large (albeit continuous) variation of the thermo-physical properties…
… may result in code execution failures.
Slide 30
SODIUM-COOLED FAST REACTOR (SFR)SODIUM-COOLED FAST REACTOR (SFR)
Characteristics• Sodium coolant, 550°C Tout• 150 to 1500 MWe• Pool or loop plant configuration
• Intermediate heat transport system
• U-TRU oxide or metal-alloy fuel
• Hexagonal assemblies of fuel pins on triangular pitch
Slide 31
SFR SAFETY R&D NEEDSSFR SAFETY R&D NEEDS
n Demonstration of passive safety design: providing assurance that the physical phenomena and related design features relied upon to achieve passive safety are adequately characterized
• Axial fuel expansion and radial core expansion– Experimental data plus deterministic models required for accurate core
representation (particularly, minor-actinide-bearing fuels)– Reduce uncertainties in T-H quantities by using more detailed models
- Multi-pin subassembly, full assembly-by-assembly, coupled neutronics-thermal-hydraulic calculation
- Accurate duct-wall and load pad temperatures required for calculating bending moments in each subassembly to characterize core restraint and expansion
- CFD tools for benchmark calculations or routine design calculations?
• Self-activated shutdown systems• Passive decay heat removal systems
– CFD models useful for resolution of complex natural circulation flow paths
Slide 32
SFR SAFETY R&D NEEDS (CONT’D)SFR SAFETY R&D NEEDS (CONT’D)
n Accommodation of extremely low probability but higher consequence accident scenarios• Demonstrate that passive mechanisms exist to preclude recriticality in
a damaged reactor• Show that debris from fuel failure is coolable within the reactor vessel
n Implication for safety analysis tools• Requires analytical and experimental investigations of mechanisms
that will ensure passively safe response to bounding events that lead to fuel damage– e.g., out-of-pile experiments involving reactor materials are
recommended for metal fuels• Local feedback and material motion modeling required
Slide 33
LEAD-COOLED FAST REACTOR (LFR)LEAD-COOLED FAST REACTOR (LFR)
Characteristics
• Pb or Pb/Bi coolant
• 550°C to 800°C outlet temperature
• Fast Spectrum
• Multi-TRU recycle
• 50–1200 MWe
• 15–30 year core life
Options• Long-life (10-30 yrs), factory-fabricated
battery (50-150 MWe) for smaller grids and developing countries
• Modular system rated at 300-400 MWe• Large monolithic plant at ~1,200 MWe• Long-term, Pb option is intended for
hydrogen generation – outlet temperature in the 750-800oC range
Slide 34
CHARACTERISTICS OF LEAD ALLOY COOLANTCHARACTERISTICS OF LEAD ALLOY COOLANT
n Low Neutron Absorption and Slowing Down Power• Allow to open the lattice, increase coolant volume fraction absent a
neutronics penalty – pumping requirements also dictate open lattice• Facilitates natural circulation
n High Boiling Temperature at Atmospheric Pressure (~1700°C)• Unpressurized primary (precludes loss of coolant accident initiator)• Margins are available to employ passive safety – based on
thermo/structural feedbacks• Potential to raise core outlet temperature (~800°C suitable for H2
production and other process heat missions)n Non-vigorous reaction with air and water
• Potential to simplify heat transport circuits• Potential to simplify refueling approaches
Slide 35
MOLTEN SALT REACTOR (MSR)MOLTEN SALT REACTOR (MSR)
HeatExchanger
Reactor
GraphiteModerator
SecondarySalt Pump
Off-gasSystem
PrimarySalt Pump
PurifiedSalt
ChemicalProcessing
Plant
Turbo-Generator
FreezePlug
Critically Safe, Passively Cooled Dump Tanks(Emergency Cooling and Shutdown)
Steam Generator
NaBF _NaFCoolant Salt
4
72LiF _Th
Fuel Salt
_BeF F _UF4 4
566Co
704Co
454Co
621Co
538 Co
Slide 36
MOLTEN SALT REACTOR (MSR)MOLTEN SALT REACTOR (MSR)
Characteristics• Molten fluoride salt fuel• 700–800°C outlet temperature• Intermediate heat transport
circuit• ~1000 MWe or larger• Low pressure (<0.5 MPa)• Graphite core structure
channels flow of actinide bearing fuel
Safety analysis issues• Modeling of nuclear, thermal,
& physio-chemical processes (e.g., FP and MA solubility, noble metal FP plate-out, …)
• Lack of established analysis capabilities
• Regulatory framework not defined